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Human Factors, Instrumentation & Controls
Improving task performance, system reliability, system and personnel safety, efficiency, and effectiveness are the division's main objectives. Its major areas of interest include task design, procedures, training, instrument and control layout and placement, stress control, anthropometrics, psychological input, and motivation.
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ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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The Standards Committee is responsible for the development and maintenance of voluntary consensus standards that address the design, analysis, and operation of components, systems, and facilities related to the application of nuclear science and technology. Find out What’s New, check out the Standards Store, or Get Involved today!
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Latest News
Norway’s Halden reactor takes first step toward decommissioning
The government of Norway has granted the transfer of the Halden research reactor from the Institute for Energy Technology (IFE) to the state agency Norwegian Nuclear Decommissioning (NND). The 25-MWt Halden boiling water reactor operated from 1958 to 2018 and was used in the research of nuclear fuel, reactor internals, plant procedures and monitoring, and human factors.
Constantine P. Tzanos, B. Dionne
Nuclear Technology | Volume 176 | Number 1 | October 2011 | Pages 93-105
Thermal Hydraulics | doi.org/10.13182/NT11-A12545
Articles are hosted by Taylor and Francis Online.
The simulation of the BR2 test A/400/1 was undertaken to support the safety analysis of the conversion of the BR2 research reactor to low-enriched uranium (LEU) fuel and to extend the validation basis of the RELAP code for analysis of the conversion of research reactors from highly enriched fuel to LEU. This test was characterized by a steady-state peak heat flux of 400 W/cm2 , total loss of flow without loss of system pressure, reactor scram, flow reversal, and reactor cooling by natural convection. This paper presents the RELAP analysis of test A/400/1 and the comparison of code predictions with experimental measurements of peak cladding temperatures during the transient at different axial locations in an instrumented fuel assembly. The simulations show that accurate representation of the pump coastdown characteristics and of the power distribution, especially after reactor scram, between the fuel assemblies and the moderator/reflector regions are critical for correct prediction of the peak cladding temperatures during the transient. Detailed MCNP and ORIGEN simulations were performed to compute the power distribution between the fuel assemblies and the moderator/reflector regions. With these distributions, the predicted peak cladding temperatures were in a good agreement with experimental measurements.
